ML23059A210
| ML23059A210 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 05/01/2023 |
| From: | Robert Lewis NRC Region 4 |
| To: | Diya F Ameren Missouri |
| Josey J | |
| References | |
| EA-22-121 IR 2022010 | |
| Download: ML23059A210 (7) | |
Text
May 01, 2023 EA-22-121 Fadi Diya, Senior Vice President and Chief Nuclear Officer Ameren Missouri 8315 County Road 459 Steedman, MO 65077
SUBJECT:
CALLAWAY PLANT - NRC INSPECTION REPORT 05000483/2022010, DISPUTED NON-CITED VIOLATION UPHELD
Dear Fadi Diya:
On October 12, 2022, the U.S. Nuclear Regulatory Commission (NRC) issued the subject report, Agencywide Documents Access and Management System (ADAMS) Accession No. ML22277A822. The inspection report documented a non-cited violation (NCV) for the failure to perform required inservice testing of residual heat removal heat exchanger pneumatically operated outlet and bypass valves (NCV 05000483/2022010-03).
In a letter dated November 14, 2022, you provided a written response and denied NCV 05000483/2022010-03, (ML22318A188 with enclosures ML22318A189 and ML22318A190). On November 21, 2022, the NRC acknowledged receipt of your letter, ML22325A336.
The NRC conducted a detailed review of your November 14, 2022, letter and examined circumstances and applicable regulatory requirements in accordance with Part I, Section 2.8 of the NRC Enforcement Manual. This review was performed by a staff member who was not involved in the original inspection effort. After consideration of the bases for your dispute of the NCV, the NRC concluded that the inspection report correctly characterizes the performance deficiency and is upholding the NCV. The details of the NRCs evaluation are documented in the enclosure.
In accordance with 10 CFR 2.390 of the NRCs Agency Rules of Practice and Procedure, a copy of this letter and its enclosure will be made available electronically for public inspection in the NRC Public Document Room and from the NRCs ADAMS, accessible from the NRC website at http://www.nrc.gov/reading-rm/adams.html.
F. Diya 2 If you have any questions concerning this matter, please contact Dr. Dustin Reinert, of my staff at 817-200-1534.
Sincerely, Signed by Lewis, Robert on 05/01/23 Robert J. Lewis Regional Administrator (Acting)
Region IV Docket No. 050000483 License No. NPF-30
Enclosure:
As stated
ML23059A210 SUNSI Review: ADAMS: Non-Publicly Available Non-Sensitive Keyword:
By: JGK Yes No Publicly Available Sensitive OFFICE SRI:EB1 SES:ACES TL:IPAT C:PBB ATL:ACES RC NAME DReinert JKramer FRamirez Munoz GWerner RKumana DCylkowski SIGNATURE /RA/ E /RA/ E /RA/ E /RA/ E /RA/ E /RA/ E DATE 02/28/23 03/03/23 03/02/23 03/01/23 03/06/23 03/07/23 OFFICE NRR OE D:DORS RA NAME RFelts DJones RLantz RLewis SIGNATURE /RA/ E /RA/ E /RA/ E /RA/ E DATE 04/07/23 04/26/23 04/26/23 05/01/23 NRC Evaluation of Licensee Response to a Non-Cited Violation (NCV)
Restatement of NCV 05000483/2022010-03 Title 10 CFR 50.55a(f), Preservice and inservice testing requirements, paragraph (4) requires, in part, that pumps and valves that are within the scope of the American Society of Mechanical Engineers Operation and Maintenance (ASME OM) Code must meet the inservice test requirements set forth in ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) of this section and that are incorporated by reference in paragraph (a)(1)(iv) of this section. Furthermore, paragraph (f)(4)(ii) requires, [i]nservice tests to verify operational readiness of pumps and valves, whose function is required for safety, conducted during successive 120-month intervals must comply with the requirements of the latest edition and addenda of the ASME OM Code incorporated by reference in paragraph (a)(1)(iv) of this section 18 months before the start of the 120-month interval. The current ASME OM Code of record for Callaway is the 2004 Edition through the 2006 Addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants, as incorporated by reference in 10 CFR 50.55a.
ASME OM Code, Subsection ISTC, paragraph ISTC-5131(a), requires that active pneumatically operated valves shall have stroke times measured when exercised in accordance with ISTC-3500.
Contrary to the above, since 2004, the licensee failed to test four active pneumatically operated ASME OM Code Class 2 valves and measure stroke times in accordance with the ASME OM Code. Specifically, the licensee failed to measure stroke times for train A and B residual heat removal (RHR) heat exchanger outlet and bypass valves in accordance with ASME OM Code, Subsection ISTC, paragraph ISTC-5131(a), as incorporated by reference in 10 CFR 50.55a.
Summary of Licensee Response In its November 14, 2022, letter, Ameren Missouri (licensee) denied that a violation of NRC requirements occurred and provided its position that Callaway was in compliance with regulatory requirements for conducting inservice testing (IST) of its RHR heat exchanger pneumatically operated outlet and bypass valves.
The licensee asserted that the subject valves are maintained in their safety position during the operational modes in which they are required to be capable of mitigating the consequences of Callaway's analyzed design basis loss of coolant accident (LOCA). Callaways design basis LOCA analysis of record assumes the plant is operating in Mode 1 at full power. The licensee claimed that there is no regulatory requirement for licensees to perform a LOCA analysis in Mode 4.
In its letter, the licensee acknowledged the existing references within the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications Bases to a Mode 4 LOCA. However, the licensee stated that these references are conflicting or incorrect and that the Bases for Technical Specification 3.5.3, ECCS - Shutdown, will be revised to delete all discussions regarding the mitigation of a LOCA or other design basis accidents. The licensee also indicated that references to WCAP-12476, Revision 1, Evaluation of LOCA During Mode 3 and 4 Operation for Westinghouse NSSS, will also be removed from Chapter 15 of the UFSAR.
Enclosure
The licensee asserted that its safe shutdown licensing basis is hot standby, and the safe shutdown design basis is cold shutdown and that the four subject valves are not required for cold shutdown operations. Finally, the licensee also indicated that its IST program has been submitted to the NRC for review multiple times and the four subject valves have been classified as passive since the initial development of the IST program.
Licensee Conclusion The licensee disagreed with NCV 05000483/2022010-03. Furthermore, the licensee believed that changing the valves ASME OM Code classification to active is inconsistent with Callaways historical and current licensing basis and that there is no regulatory requirement for licensees to perform a LOCA analysis in Mode 4.
NRC Evaluation The NRC staff performed an independent review of the licensees position as described in its letter dated November 15, 2022, for NCV 05000483/2022010-03. The NRC staff reviewed:
Regulatory requirements in 10 CFR 50.46, 10 CFR Part 50 Appendix K, the ASME OM Code, and Callaways Technical Specifications, Design basis information in Callaways UFSAR, and Guidance in the Westinghouse Standard Technical Specifications Bases and NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Revision 3.
Title 10 CFR 50.46(a)(1)(i) requires, in part, that emergency core cooling system (ECCS) cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated. Although Appendix K to 10 CFR Part 50 requires that the licensee select evaluation model inputs that result in the most severe calculated consequences for the spectrum of postulated breaks and single failures that are analyzed, the ECCS system must provide protection in all modes of operation as specified in 10 CFR 50.46(a)(1)(i). As such, NUREG-0800, Section 15.6.5, Loss-of-Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, instructs the reviewer of the ECCS performance analysis to confirm that the parameters and assumptions used for the calculations were conservatively chosen. In addition to the reviews of the analyses provided in Chapter 15 of the SAR, NUREG-0800 Section 6.3, Emergency Core Cooling System, instructs the reviewer to review the ECCS response in LOCAs to confirm the ECCS design adequacy for all modes of reactor operation (e.g., full power, low power, hot standby, cold shutdown, partial loop isolation).
The NRC staff considered the licensees assertion that the Bases for Technical Specification 3.5.3, ECCS - Shutdown, contain conflicting, misleading, and incorrect information regarding the applicable safety analyses and single failure criteria associated with ECCS in Mode 4. The Westinghouse Standard Technical Specifications Bases statement that The Applicable Safety Analyses section of Bases 3.5.2 also applies to this Bases section, is appropriate for inclusion in the Bases for Technical Specification 3.5.3 because it is assumed 2
that a postulated design basis accident (DBA) at full power conditions would bound DBAs occurring in lower modes. Additionally, the NRC staff noted that the plant specific Technical Specification Bases does not contain this reference to Specification 3.5.2.
Furthermore, the licensees claim that it cannot be assumed that the Mode 1 LOCA analysis is bounding of Mode 4 conditions, since the safety injection (SI) on Containment Pressure - High 1 and Pressurizer Pressure - Low ESFAS functions, as well as the SI accumulators, which are assumed to function in the LOCA analysis that is performed at 102 percent of the licensed power level, are not available in Mode 4, has been accounted for within the scope of the requirements of Technical Specification 3.5.3. The Bases state that, It is understood in these reductions that certain automatic safety injection (SI) actuation is not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a DBA.
The licensee asserted that the Technical Specification 3.5.3 requirement for only one operable train of RHR in Mode 4 conflicts with the single failure requirement of General Design Criterion 35. The staff noted that this also has been accounted for within the scope of Technical Specification 3.5.3. The Technical Specification Bases do clearly acknowledge that due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (DBA), the ECCS operational requirements are reduced.
Thus, during this mode of operation, single failures of ECCS are not considered. Additionally, Standard Technical Specification Change Traveler 90, a joint owners group and NRC approved document which was incorporated into the licensees Technical Specification in License Amendment 133, recognized that single failures are not considered during Mode 4 operations and that the Residual Heat Removal system was required to function in the ECCS mode while in Mode 4.
The staff acknowledges the licensees statement that there is no regulatory requirement for licensees to perform a LOCA analysis in Mode 4. However, the Mode 4 DBA may be considered bounded by the full power LOCA analysis. This appears to be consistent with the licensees historical interpretation of its licensing basis. The staff reviewed Callaway Licensee Event Report 2010-001-01 which was submitted to the NRC for the failure to fully implement the Mode 4 LOCA mitigation capability required by Technical Specification 3.5.3 within site procedures. The licensees Corrective Action to prevent recurrence was to develop and implement a site-specific MODE 4 LOCA flow evaluation based on the WCAP-12476 analysis to confirm the ECCS requirements for accident mitigation. The licensee subsequently incorporated Westinghouse letter SCP-10-31, Transmittal of Mode 4 Small Break LOCA (SBLOCA) RHR Flow Evaluation for Callaway (SCP) - Phase 3 - Revision 1, into its Technical Specification Bases.
The staff also considered the licensees position regarding the treatment of the four subject valves within the IST program. The staff noted that the Callaway UFSAR indicates a design basis capability of the valves to achieve cold shutdown. The use of the subject valves in the RHR system is the normal process for achieving cold shutdown. Even though there may be alternate success paths for achieving cold shutdown such as throttling component cooling water flow or manually controlling RHR pumps such that these valves are not required for cold operations, the subject air-operated valves still have an active safety function to change obturator position to achieve cold shutdown. The ASME OM Code beginning with the 1995 edition specifies scope using the same language as 10 CFR 50.2 for safety-related components. The ASME OM code does not have different requirements based upon operating mode. Therefore, the existence of a Mode 4 LOCA analysis is not relevant to the question of the 3
IST scope for the subject valves. Additionally, the staff noted that 10 CFR 50.55a requires licensees to update the IST programs every ten years.
NRC Conclusion The NRC staff concludes that the NCV as documented in the October 12, 2022, inspection report remains valid. The licensee failed to test four active pneumatically operated ASME OM Code Class 2 valves and measure stroke times in accordance with the ASME OM Code.
Therefore, the NRC is upholding NCV 05000483/2022010-03, Failure to Perform Required Inservice Testing of Residual Heat Removal Heat Exchanger Pneumatically (Air) Operated Outlet and Bypass Valves.
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